1. The thermal-hydraulics of a boiling water nuclear reactor;Lahey,1979
2. Comparison of Assert subchannel code with Marviken bundle data;Tahir;AECL Report, AECL-8352,1984
3. Coolant interchannel mixing in reactor fuel rod bundle, single-phase coolants;Rogers,1968
4. Two-phase flow mixing in rod bundle subchannels;Gonzalez-Santalo,1972
5. COBRA III-C: a digital computer program for steady state and transient thermal analysis of rod bundle nuclear fuel elements;Rowe;BNWL-1695,1973